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RISS 인기검색어
- 검색키워드
- 저자 : Mutegi Peter Mutembei (검색결과 4 건)
- 무료
- 기관 내 무료
- 유료
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Design Re-engineering of APR1400 Reactor Vessel
Mutegi Peter Mutembei,Ihn Namgung 대한기계학회 2016 대한기계학회 춘추학술대회 Vol.2016 No.12
This paper describes the conceptual re-engineering of the RV (Reactor Vessel) for the APR1400. The RV is classified as safety class 1 and therefore must adhere strictly to the rules of ASME BPVC section III, subsection NB and seismic category I. This study explores designing the RV by following the ASME guidelines and making a comparative study with the current design. Focus is on the determining the design parameters of RV major nozzles and their reinforcement, given that they present structural discontinuities that may weaken the nozzle opening if not compensated for adequately. The design parameters that are not derived from ASME code guidelines were referred from the APR1400 SSAR. The obtained parameters offered an input to creating a 3D model using the CATIA software 3D platform. A comparative study was done on the ASME Code specification RV geometrical model and the standard geometrical model.
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Design Verification of APR1400 Reactor Vessel Through Re-engineering Approach
Mutembei, Mutegi Peter,Namgung, Ihn The Korean Society of Systems Engineering 2017 시스템엔지니어링학술지 Vol.13 No.1
This paper describes verification of APR1400 reactor vessel by applying the system engineering approach, in which the design re-engineering method is used to check the design parameters of APR1400 RV (reactor vessel). The RV is classified as safety class 1 and therefore must adhere strictly to the rules of ASME BPVC section III, subsection NB and seismic category I. This study explores designing the RV by following the ASME guidelines and making a comparative study with the current design. To meet this objective we apply system engineering methodologies to structure the process and allow for verification and validation of the major RV design parameters such as thickness of RV. The structural thicknesses of various part of RV are determined as well as reinforcements on the RV major nozzles. A 3D virtual reality model was created based on the design parameters using CATIA V5 and animation using Dassault Composer V2016. A comparison of re-engineered ARP1400 RV and standard APR1400 RV was done to show which design parameters were taken more conservative approach.
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Heat Transfer Analysis of PLUS7 Fuel Rod For APR1400 Using ANSYS
Sang-Jun Park(박상준),Mutegi Peter Mutembei,Ihn Namgung 대한기계학회 2016 대한기계학회 춘추학술대회 Vol.2016 No.12
The fuel rod is composed of fuel pellets, fill gas, end caps, plenum spring and cladding. The heat is transferred from the pellet outward by conduction through the pellet, fill gas, and cladding and further by convection from the cladding surface to the coolant in the flow channel. The goal of this paper is to demonstrate the temperature and heat flux change from the fuel centerline to the cladding surface when having maximum fuel centerline temperature at 100% power. This phenomenon is modelled using the ANSYS thermal analysis tool and the results compared to the standard values given APR1400 SSAR. The close proximity of the results obtained ascertains the accuracy of the FEM analysis even with the simplified 2D axisymmetric model, whose results matched the 3D model. This paper demonstrated the validity of commercial software instead of proprietary in-house code that is more costly to develop and maintenance.
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Systems Engineering Approach to the Heat Transfer Analysis of PLUS 7 Fuel Rod Using ANSYS FEM Code
Park, Sang-Jun,Mutembei, Mutegi Peter,Namgung, Ihn The Korean Society of Systems Engineering 2017 시스템엔지니어링학술지 Vol.13 No.1
This paper describes the system engineering approach for the heat transfer analysis of plus7 fuel rod for APR1400 using, a commercial software, ANSYS. The fuel rod is composed of fuel pellets, fill gas, end caps, plenum spring and cladding. The heat is transferred from the pellet outward by conduction through the pellet, fill gas and cladding and further by convection from the cladding surface to the coolant in the flow channel. The goal of this paper is to demonstrate the temperature and heat flux change from the fuel centerline to the cladding surface when having maximum fuel centerline temperature at 100% power. This phenomenon is modelled using the ANSYS FEM code and analyzed for steady state temperature distribution across the fuel pellet and clad and the results were compared to the standard values given in APR1400 SSAR. Specifically the applicability of commercial software in the evaluation of nuclear fuel temperature distribution has been accounted. It is note that special codes have been used for fuel rod mechanical analysis which calculates interrelated effects of temperature, pressure, cladding elastic and plastic behavior, fission gas release, and fuel densification and swelling under the time-varying irradiation conditions. To satisfactorily meet this objective we apply system engineering methodologies to formulate the process and allow for verification and validation of the results acquired. The close proximity of the results obtained validated the accuracy of the FEM analysis of the 2D axisymmetric model and 3D model. This result demonstrated the validity of commercial software instead of proprietary in-house code that is more costly to develop and maintain.
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